Study of a spherical tokamak based volumetric neutron source

Transcription

1 Fusion Engineering and Design 38 (1998) Study of a spherical tokamak based volumetric neutron source E.T. Cheng a, *, Y.K. Martin Peng b, Ralph Cerbone a, Paul Fogarty b, John D. Galambos b, E.A. Mogahed c, Brad Nelson b, Massoud Simnad d, Igor Sviatoslavsky c, Mark Tillack d a TSI Research, Incorporated, Solana Beach, CA , USA b Oak Ridge National Laboratory, Oak Ridge, TN , USA c Uni ersity of Wisconsin, Madison, WI , USA d Uni ersity of California, San Diego, CA , USA Received 15 August 1996; accepted 26 June 1997 Abstract With the worldwide development of fusion power focusing on the design of the International Thermonuclear Experimental Reactor (ITER), developmental strategies for the demonstration fusion power plant (DEMO) are being discussed. A relatively prudent strategy is to construct and operate a small deuterium tritium fuelled volumetric neutron source (VNS) in parallel with ITER. The VNS is to provide, over a period less than 20 years, a relatively high fusion neutron fluence of 6 MW year m 2 and wall loading of 1 MW m 2 or more, over an accessible blanket test area of more than 10 m 2. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components, materials, and their combinations for DEMO and future commercial power plants. The effort of this study has established the potential of the spherical tokamak as a credible VNS concept that satisfies the above requirements Elsevier Science S.A. Keywords: ITER; Volumetric neutron source; Spherical tokamak; DEMO 1. Introduction A fusion plasma based volumetric neutron source (VNS) concept has been identified in recent studies as a prudent intermediate step to develop the necessary technology for reactor components of future fusion power plants [1]. This VNS would provide, over a period of operation less than 20 years, a relatively high fusion neutron fluence of 6 * Corresponding author. Tel.: ; fax: ; etcheng@cts.com MW year m 2, at no less than 1 MW m 2 in fusion neutron wall loading, and over an accessible test area of more than 10 m 2. Such a VNS would complement ITER in testing, developing, and qualifying nuclear technology components, materials, and their combinations for the Pilot Plants, DEMO, and future commercial power plants. A VNS concept based on the low aspect ratio tokamak (the spherical tokamak) using a normal conducting, single-turn center leg of TF coil is an attractive candidate among all tokamak concepts /98/$ Elsevier Science S.A. All rights reserved. PII S (97)

2 220 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) The spherical tokamak (ST) based volumetric neutron source (VNS) is one of the most innovative and attractive concepts to satisfy the nuclear technology testing needs in developing fusion power. Power blankets, divertors, and first walls of relevance to power plants can be tested in the ST based VNS. The operating experience and reliability data of these components can be obtained more economically in this VNS than in a full-scale, large tokamak experimental power reactor. This VNS may offer an added opportunity to cost-effectively test issues associated with long lifetime of these components in a tokamak plasma environment. If the technical feasibility is established, an ST based VNS can be a major facility to develop the nuclear technology, materials, and operational data base needed for the design, construction, and operation of a DEMO. In this paper we report the results of an investigation which was carried out with the following objectives: (1) to identify feasible small spherical tokamak (ST) based concepts for the fusion core of a VNS facility dedicated to testing and qualifying fusion nuclear technology for blankets, first wall, and divertors; and (2) to estimate the performance characteristics of the ST based VNS facility. The bulk of this paper is divided into sections as given below: Section 2. Physics basis, VNS objectives, and divertor heat flux 2.1. Experimental status 2.2. Physics projections and next step proof-ofprinciple tests 2.3. Design concept and parameters 2.4. Thick naturally diverted SOL and modest divertor heat flux Section 3. Mechanical design considerations 3.1. Design of the center-post 3.2. Thermal-hydraulics of center-post 3.3. Structural consideration 3.4. Sliding electrical connection 3.5. First wall for center-post 3.6. The divertor design 3.7. Center-post fabrication 3.8. Center-post replacement 3.9. Other maintenance requirements Power supplies Section 4. Nuclear performance 4.1. Neutronics model 4.2. Neutron source 4.3. Neutron fluxes 4.4. Neutron wall loading 4.5. Nuclear heating 4.6. Atomic displacement and gas production rate 4.7. Copper transmutation in center-post 4.8. Decay heat 4.9. Waste disposal Test section Comparison with the 2000 MW demo-like plant Section 5. Cost of fusion core and support systems Section 6. Summary and conclusions 2. Physics basis, VNS projections, and divertor heat flux The physics basis, the projections for the design parameters, and the estimates for the divertor heat flux are presented in this section for the small VNS fusion core based on the spherical tokamak (ST). The results will provide strong justification for the scientific feasibility for such a VNS to become available within the next decade. The estimates of plasma performance and design parameters reported here provide key input to the VNS mechanical design and nuclear performance considerations to be presented in Sections 3 and 4, respectively. Section 2.1 summarizes the encouraging data from several small pioneering ST experiments. Section 2.2 summarizes the highly attractive plasma properties for fusion ST plasmas that can be projected based on these data. These projections define the physics principles to be tested in the next-step ST experiments that can be brought into operation in the next 3 years. Section 2.3 summarizes the design parameters and plasma performance calculated for the minimum cost VNS fusion cores, using a modified version of the SuperCode [2] to account for the ST plasmas and

3 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 1 Representative parameters of recent and present pioneering spherical tokamak experiments Device name R 0 (cm) A (R 0 /a) I p (ka) t p (ms) B t0 (kg) T e0 (ev) HSE+rod [6] (FRG, 1987) ROTOMAK [7] (Australia, 1987) FBX II [8] (Japan, 1990) SPHEX Tokamak [9] (UK, 1991) START [5] (UK, 1991) TS-3 [10,11] Japan, 1991) TS-3, low-a [10,11] (Japan, 1993) HIT [12] (USA, 1994) CDX-U [12] (USA, 1994) TST [13] (Japan, 1995) design features [3,4]. Section 2.4 presents an approach to ensure relatively modest heat fluxes on the plasma facing components in the VNS, based on the thick, naturally diverted scrape-off layer (SOL) observed and studied in START [5]. A factor of two margin in reducing the heat flux further will be shown to be available by increasing the plasma safety factor without increasing the fusion core size and cost significantly. Recent reviews by the US Department of Energy Fusion Energy Advisory Committee (FEAC) of the new DOE strategy for fusion research have increased the probability for building within a few years the next-step ST physics experiments in the US. This would greatly enhance the likelihood for establishing within the next decade the data base needed to design and build a highly cost-effective VNS based on the ST configuration Experimental status Spherical tokamak experimental research began about 9 years ago in Table 1 lists recently operating pioneering ST experiments and their major parameters [6 13]. It is seen that the experiments have been limited to mostly ohmic plasmas of modest sizes, currents, and durations. Despite such limitations, the data obtained have been very encouraging in essentially all physics issues of importance to future economic fusion energy applications. The measurements in cases of significant plasma temperatures (most notably in START, HIT, CDX-U, TS-3, etc.) have been consistently better than the projections based on the standard tokamak data base. Listed in Table 2 are the plasma physics features of importance to the fusion core for future power applications. These features are grouped into the following desirable plasma characteristics: 1. High power density in small volume requiring only low external magnetic field, 2. Reliable operation, 3. Low recirculating power, 4. Successful power, particle, and impurity handling, and 5. Low power to start up current and fusion burn. Measurements in START [5,14] have indicated that the plasma energy confinement time is about twice those suggested by most scaling expressions based on the standard tokamak data. It is appropriate to use the H-factor (H f ) relative to the so-called ITER-89P scaling [15] as a measure. An average beta ( t ) up to 2% has been observed in START with ohmic heating alone, leading to a normalized beta ( Nt ) of about 1.5% m T MA 1. The Shaping Factor (S), first identified by Lazarus [16] to indicate the best performance plasmas in DIII-D with S up to about 8 MA m 1 T 1, is measured to be as high as 20 MA m 1 T 1 in START. On the other hand, a more recent measurement on DIII-D [17] showed that it is possible to achieve H-mode confinement in plasmas limited only at the inboard. This suggests that H-mode confinement should also be possible

4 222 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 2 Important physics features, present data, test goals for next-step experiments, and desirable conditions for ST fusion cores of VNS and future power plants Important physics features Present data MA test goals VNS fusion core Power plants 1a. High energy confinement factors H f b. High stable t (%)/ Nt (% m T MA 1 ) 2/1.5 45/8.6 45/6 50/9 1c. Strong shaping S=I p q 95 /ab t0 (MA m 1 T 1 ) a. High =b/a for vertical stability b. High H e Nt for infrequent or no disruptions a. High pressure-driven current fraction I pres /I p b. High current drive coef. CD (10 19 A m 1 T 1 ) No data c. High toroidal field utilization I p /I tfc a. Thick outboard pressure e-folding p-sol (cm) b. Small inboard exhaust power fraction, f power c. Long divertor channel (m) a. High I p (MA) via noninductive startup b. Moderate auxiliary heating power (MW) in similarly limited ST plasmas. These results indicate the potential for high confinement and in the ST plasma. For hollow current profiles, stable vertical elongations greater than 3 have been observed in START without strong externally applied plasma shaping. So far not a single low aspect ratio plasma in START has ended in disruption, despite occasional strong internal magnetic field line reconnections, which are frequently observed in the standard tokamak. These results indicate the potential for robust stability of the ST plasma. A bootstrap fraction of 0.2 has been inferred for the ohmic plasmas in START. For aspect ratios as low as 1.1 in TS-3 [11], a ratio of I p /I tfc as high as three has been shown to maintain global stability of the ST plasma. These results indicate the potential for low recirculating power in future ST fusion applications. Neutral beam heating experiments [18], planned for 1996 for START, should provide initial data for non-inductive current drive in the ST plasma. A thick outboard SOL has been observed in the inboard-limited, naturally elongated, and naturally diverted plasmas in START, without poloidal divertor x-points at the plasma edge. (Probe measurements [19] have shown pressure SOL thickness in the range of 2 3 cm for such plasmas, which is consistent with the criterion of marginal stability for MHD pressure driven modes [4]). These may be features unique to the ST plasma. Detailed measurements in START [19] have shown that no more than 10% of the total plasma heating power reached the inboard limiter in such cases. A thick diverted SOL channel as long as 50 cm is also obtained naturally in these plasmas. These results indicate the unique potential for much lowered limiter and divertor heat fluxes in future ST fusion plasmas. Coaxial helicity injection (CHI) experiments in HIT [12] have succeeded in producing ST plasmas of currents up to 250 ka with relatively high energy efficiency, in the absence of induction from the solenoid coils commonly used in the standard tokamaks. This result indicates the potential for successful noninductive start up of full plasma current in future ST applications. Initial neutral beam injection heating results are anticipated from START [18] in Physics projections and next step proof-of-principle tests These new ST data have shed light on the ST plasma physics and led to rapid progress in theoretical analyses of possible ST plasma properties.

5 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 3 Representative parameters for next step ST experiments under construction or review Device name Status R 0 (m) A (R 0 /a) I p (MA) t p (s) B t0 (kg) Auxiliary power (MW) GLOBUS-M [26] (Russia) Construction USTX [27] (USA) Review MAST [28] (UK) Design NSTX [29] (USA) Design Recent calculations of highly attractive ST plasma properties are summarized in Table 2, and were presented by researchers from PPPL, GA, ORNL etc. at recent international ST workshops [20,21] and plasma physics conferences. For example, MHD stable average betas ( t ) in the range of 50% producing order-unity fractions of self-driven bootstrap currents have been calculated for naturally elongated ST plasmas [22,23]. Strong stabilization of tokamak microinstabilities, commonly believed to be responsible for the anomalously large energy losses from tokamak plasmas, has been calculated for the ST plasma [24]. Extrapolation of the naturally diverted thick SOL to ST plasmas of high power densities indicates relatively modest heat fluxes on the limiters and the divertor plates [25]. These exciting projections have led to several proposals in the world to carry out physics tests at the MA level, where high-temperature collisionless ST plasmas are anticipated [20]. Recent examples for experiments under construction, being designed, or under review are listed in Table 3 [26 29]. It is seen that these experiments are of modest size, cost, and engineering requirements, and can utilize equipment and facilities presently available in the fusion community. The combination of complementary designs such as MAST and NSTX ensures our ability to address essentially all physics issues of interest to the VNS, except those related directly to the use of tritium. The validity of the physics basis for these experiments was affirmed in recent physics reviews [30]. Such experiments, if built in a few years, are expected to provide in the next decade the data needed to design and build small, low-cost VNS s based on the ST configuration Design concept and parameters The ST physics assumptions for a near-term D T-fuelled VNS, consistent with the latest experimental data and theoretical analyses just discussed, are listed in Table 4. Also listed are the plasma properties that can be reasonably assumed for ST power plants years in the future. Also listed are the maximum plasma properties that can be reasonably assumed for ST power plants years in the future. It is seen that the plasma properties suggested for the VNS are often significantly less than these maximum possible values. Successful operation of the VNS plasma is therefore more likely in the nearer term than power plants. The basic design configuration for the VNS fusion core is depicted in Fig. 1. Detailed discussion of the mechanical design concept will be presented in Section 3. Important design features that influence the VNS parameters include: 1. A demountable, single-turn, water-cooled copper (dispersion strengthened copper, GlidCop) conductor for the center leg of the TF coil, 2. Space for actively cooled limiter tiles on the center leg where it is exposed to the plasma, 3. Inboard-limited, outboard-diverted, and naturally elongated ST plasma configuration with thick outboard SOL, 4. Top-down symmetric divertors with long channels, 5. Outboard access for nuclear test modules, 6. Tangential neutral beam injection heating and current drive, 7. Nuclear shielding adequate for hands-on access outside of shielding after shutdown, 8. Poloidal field coils inboard of the TF coil,

6 224 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 1. Elevation view depicting the major design features of a VNS based on the spherical tokamak configuration. 9. Combined TF coil return leg and vacuum vessel to simplify load path and maintenance access. The physics and engineering models for the SuperCode [2] have been modified to account for these design features [4]. Calculations are carried out for VNS designs that satisfy all the physics design constraints provided in Table 2, while adhering to the engineering design constraints stemming from the above list (see Section 3 for detail). Two VNS cases are presented in Table 4 below. Case I represents the minimum cost design that provides an average fusion neutron wall load of 1 MW m 2 over an outboard accessible wall area of no less than 14 m 2. This accommodates a minimum outboard wall area of 10 m 2 for blanket test modules and allows adequate access for plasma heating and current drive. Case II is obtained by using the Case I device while maximizing the neutron wall load by adjusting the plasma operating conditions and heating power, regardless of additional cost. These results are similar to the basic concepts for the small VNS presented earlier [31]. The main parameters for TFTR and JET, the largest D T-fuelled tokamak experiments to date, are provided for comparison.

7 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 4 Design parameters for VNS based on the ST configuration in contrast with TFTR and JET VNS (I-II) TFTR JET Major radius, R 0 (m) Minor radius, a (m) Aspect ratio, R 0 /a Plasma elongation, (=b/a) Plasma current, I p (MA) Applied toroidal field at R 0, B t0 (T) Nominal edge safety factor, q Average density, n e (10 20 m 3 ) Peak ion temperature, T i0 (kev) Heating and current drive power (MW) Fusion power (MW) a 20 Duration of D T burn (s) s.s. 2 5 Plasma surface area (m 2 ) Total energy flux at plasma edge, edge (MW m 2 ) Heating power to R 0 ratio (MW m 1 ) Average neutron wall load (MW m 2 ) Resistive power for TF coil (MW) Achieved in It is seen that, relative to TFTR and JET, the VNS fusion core is small in size (R 0 =0.8 m, a=0.6 m), toroidal field (1.8 T), plasma surface area (29 m 2 ), and resistive power for the TF coil (100 MW). It is comparable in plasma heating power (21 54 MW) and ion temperature (24 kev). It is higher in plasma density ( m 3 ), plasma current ( MA), edge safety factor ( ), fusion power (39 59 MW), neutron wall load (1 1.5 MW m 2 ). However, it is much higher in total energy flux at the plasma surface (1 2.3 MW m 2 ) and ratio of heating power to R 0 (36 82MWm 1 ). These comparisons suggest that a VNS based on the ST configuration has a plasma performance comparable to a standard tokamak of low energy amplification (Q=1 2) in all areas except the very high heat fluxes at the divertor. In the next section we will show that the divertor heat fluxes for an ST-based VNS can be less than those expected of tokamaks, based on recent progress in experimental and theoretical studies of the SOL plasma in the ST Thick naturally di erted SOL and modest di ertor heat flux For densities significantly below the density limit, high heat flux at the plasma surface lead to high heat flux at the divertor plate if the SOL thickness remains unchanged by changes in heat flux. This has been observed to be largely the case (with p 1 cm) for H-mode plasmas in standard divertor tokamaks of varying sizes and heating power. However, recent measurements in START and theoretical progress show that this characteristics for the SOL in the standard tokamak does not apply in the ST. Progress of importance to this issue can be summarized as follows: (1) p for the outboard SOL in START has been measured [5] to vary by an order of magnitude when the field line connection length of the SOL (L conn ) is varied by a factor of about three. (2) This is divergent from the predictions assuming transport via thermal and particle diffusion, commonly used in standard tokamak studies [32], but consistent with the predictions assuming marginal MHD stability for the SOL plasma [4]. The latter model suggests that,

8 226 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) p 0.02 (n i T i )R 0 ( f c q 95 /B t0 ) 2, [m, m 3, kev, T] 2 which is proportional to L conn. Here p is the pressure e-folding thickness, n i and T i are the plasma density and temperature at the plasma edge, respectively, and f c equals 1 for up-down symmetric divertors and 2 for the single-null divertor. (3) Recent probe measurements in START [19] showed that about 10% of the total plasma heat flux at the plasma edge reaches the inboard limiter for ST plasmas with natural divertors, a natural elongation of 1.5, and an aspect ratio of 1.4. This suggests that only minimal heat fluxes will reach the inboard limiter in ST plasmas with lower aspect ratio and larger elongation [33]. (4) Recent measurements in DIII-D [34,35] suggested that for H-mode plasmas, the following condition relating the plasma parameters to the total energy flux ( edge ) at the plasma edge is roughly satisfied: n i T i 2n e T e 0.8 edge. [10 20 m 3, kev, MW m 2 ] The preceding results can be combined to show that the outboard divertor heat flux ( div ) for the ST, averaged over the foot print of one pressure e-folding thickness, can be approximately given by: div 200[(1 f rad )/f R f exp f 2 c] 1/2 (B t0 /q 95 ) 2. [MW m 2,T] Here (1 f rad ) is the fraction of plasma energy that reaches the divertor, R div = f R R 0 is the mean radius of the divertor plate, f exp is the average factor of area expansion of the divertor flux tube beyond the mid-plane area, and is the inverse aspect ratio. It is seen that the divertor plate heat flux ( div ) is roughly independent of the edge energy flux ( edge ) for the ST plasma, depends weakly on the plasma geometry (, ), but depends strongly on the ratio of B t0 /q 95. Assuming the VNS cases in Table 4 with B t0 =1.8 T, q 95 = 5 5.7, f rad =0.5, f R =0.4, f exp =10, f c =1, = 0.75, and =2.3, div is found to be roughly MW m 2. However, for a standard tokamak with B t0 =5.7 T, q 95 =3.2, f rad =0.5, f R = 0.8, f exp =5, f c =2, =0.37, and =1.6, div would increase to about 9.3 MW m 2, which is roughly consistent with the estimates commonly obtained for designs such as ITER [32]. For conservatism, we do not rely on gas target to reduce the divertor heat flux and assume f rad =0.5. It is seen that the combination of natural elongation, natural divertor, and MHD-instability driven thickness for the SOL have the potential for reducing the divertor heat flux for the ST plasma to levels much below that for the standard tokamak of comparable neutron wall load. Also, ST power plants have been estimated [4,22] to prefer q 95 as high as 20 to enhance the bootstrap fraction of the plasma current. Even lower average heat fluxes are therefore anticipated for the ST power plants than that presently estimated for the VNS. To introduce additional margins in lowering the divertor heat flux for the VNS, we exploit the strong q 95 dependence projected for div. A sequence of minimum size VNS designs are calculated using the SuperCode by requiring that q 95 increases to as high as 10. The results are plotted in Fig. 2 below. It is seen that the plasma major radius, heating power, and fusion power do not increase significantly until q In the meantime, the average divertor heat flux div can decrease by about a factor of 2 while p increases from about 10 cm to about 20 cm. The strong q 95 -dependence therefore lends itself to providing margins in lowering the divertor heat flux for the ST-based VNS. Effects due to possible disruptions were not considered in this initial study. At present, the limited experimental results (START, CDX-U) suggest that the ST plasmas may be resilient to disruptions and produce reduced halo currents when the plasma edge safety factors are high ( 5). The disruption effects should be considered, however, when improved understanding is obtained in the ST physics research program (NSTX, MAST). The next-step ST experiments at the MA level will afford in near term opportunities to verify these very interesting SOL properties for high temperature ST plasmas.

9 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Mechanical design considerations A challenging aspect of an ST (spherical torus) VNS (volumetric neutron source) is the design of the copper center-post. The center-post has a major role in the viability of this concept. All the currents from the outboard TF (toroidal field) coil legs converge at the center-post, and the total current flows through it to reconnect again with the outboard TF coil legs on the other end. This very large current, passing through a relatively thin conductor provides a design challenge with respect to cooling, mechanical forces and fabrication. An elevation view of the ST-VNS is shown in Fig. 1. The overall shape is that of a bell jar connecting to a center-post running through the center. The so called bell jar constitutes the outboard TF coil return legs, and at the same time forms the vacuum boundary for the device. On the inside it contains the fusion core, consisting of the plasma volume, FW (first wall), divertors, test modules, neutral beams and shield. A 1.8 m person is shown for size perspective. There are 3 PF (poloidal field) coils on the upper and lower halves of the device. These coils will be either superconducting or normal (Cu or Al) operated at cryogenic temperatures. Divertors are shown on the upper and lower extremities of the D shaped plasma. Although pump-out ports do not appear in the figure, they would be distributed on the outer perimeter of the divertor plates, with appropriate bends to inhibit neutron streaming. Fig. 3 is a top view of the device, showing two neutral beams and six test module spaces. Vacuum pumps are located between test module spaces and neutral beams. One such pump port is shown next to the standing person in Fig. 1. Water coolant connections are shown supplying and emptying coolant to the center-post, the divertor modules, the outboard TF coils and the shield. The design of VNS is specially geared for quick and easy replacement of radiation vulnerable components, such as the center-post, divertor plates and outboard FW sections. By removing the large flanges on the top and bottom of the bell jar, and sliding out the center-post and divertor plugs, the device becomes easily accessible for replacing outboard shield FW segments. One of the concerns is the use of water cooling in the VNS when there will be blanket test modules containing liquid lithium. At the present time an approach has been considered that whenever water cooling will be used, there will be double containment. Thus, two barriers will have to be breached before the potential of water coming in contact with lithium will occur. Further evaluation of this approach will be made in future work Design of the center-post Fig. 2. The effects of changing q 95 on the major parameters for VNS based on the ST configuration. The center-post in VNS is a long thin water cooled copper column 16.7 cm in radius at midplane and 32 cm radius at the extremities. For

10 228 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 3. Top view of the ST VNS. purposes of analysis, the center-post has been divided into five segments as shown in Fig. 4. Cross sections of 1/12 of the center-post are shown enlarged but in the same relative scale at three locations, at mid-plane, midway in the upper transition section and at the upper straight section. The same cross section would exist midway in the lower transition section and the lower straight section. Fig. 4. Copper center-post geometry. The cross sections depicted in Fig. 4 show a very uniform coolant distribution at all points along the length of the center-post. The bulk of the heating in the center-post is ohmic, arising from the current, and is distributed uniformly over the cross sections. A small amount ( 2%) is nuclear heating which is concentrated near the outer surface of the center-post, but contributes negligibly to the total heating. The cross sections in Fig. 4 show eight rows of holes distributed on equal radial distances. Each row has an increment of one hole relative to the previous row, ending up with 36 holes in each of the 1/12 segments for a total of 432 holes. These holes represent coolant tubes that extend the whole length of the center-post. The diameter of the tubes changes along the length of the centerpost, as does the location. The same tube tracks the length of the center-post but its diameter and radial location changes from one level to the next. Consequently, the coolant fraction at each section is different, with a maximum value of 38% at mid-plane and dropping to 19% at the extremities. The reasoning behind these coolant fractions will be evident from the thermal hydraulics section. Table 5 gives the physical parameters of the center-post.

11 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 5 Physical parameters of center-post Upper straight Upper transition Middle straight Lower transition Lower straight Radius (cm) Area (cm 2 ) Length (cm) Coolant fraction (%) Number of tubes Diameter of tubes (cm) Area tube (cm 2 ) Perimeter of tubes (cm) Thermal hydraulics of center-post In the reference case analyzed, the current in the center-post is 7.16 MA and the Cu conducting area at mid-plane is 543 cm 2 giving a maximum current density of ka cm 2. The high ohmic heating at mid-plane requires a large fraction of coolant to minimize the Cu temperature. Since the electrical resistivity of Cu increases with temperature, it is important to keep the temperature low and as uniform as possible. The thermal hydraulic analysis has been performed using Glidcop (Cu alloy), a patented alloy produced by SCM Corporation [36] using powder metallurgy. This alloy has a high electrical as well as thermal conductivities while maintaining high strength at elevated temperatures. Table 6 gives the properties of Glidcop AL-25 as a function of temperature and Fig. 5 gives the electrical resistivity of Glidcop AL-15, a formulation which has only 0.15% of Al 2 O 3 as compared with 0.25% for the AL-25. A fit to the curve is given at the top of the figure. The yield and ultimate strength value for AL-15 are typically 4% lower than AL-25 while the electrical conductivity is higher by 7.5% and the thermal conductivity is higher by 6%. The analysis has been made using the values for AL- 15. The center-post is cooled with water flowing through the tubes from top to bottom. The water enters the center-post at 40 C and exists at 118 C. Because the diameter of the tubes changes while the mass flow rate is constant, the velocity changes along the length of the center-post. The velocity is the highest in the central straight section, and at the start of operation before the resistivity increases from nuclear transmutation, it is 7.35 m s 1. At the extremities, that is, in the upper and lower straight sections, it is 4.0 m s 1. Table 7 gives the thermal hydraulic parameters of the center-post at the start of operation. The Table is divided into the same 5 sections as indicated in Fig. 4. The water mass flow rate is the same for all, 245 kg s 1, the inlet and outlet temperatures from each zone are indicated, the water velocity, Glidcop electric resistivity and ohmic powers are listed. Two MW of nuclear heating are added to the total ohmic heating with a uniform vertical distribution in the middle section. Radial distribution was neglected because of the very high conductivity of copper and because nuclear heating constitutes only a small fraction of the total heating. Further, the heat flux to the tubes, heat transfer coefficients, Glidcop thermal conductivity and Glidcop maximum temperatures are shown. Finally, the water pressure drop is shown for each section, adding up to a total pressure drop of MPa. The pumping power is 41.2 kw. Fig. 6 gives the percentage increment in the Glidcop electric resistivity due to neutron induced transmutation after 6 full-power years (FPY) of operation (see Section 4.7). The average increase integrated vertically and radially, and normalized to the increase in area with radius is 39% for the middle straight section and 18.5% for the upper and lower transition sections. To make up for increased ohmic heating due to resistivity increase, the water mass flow rate is increased. The temperature rise in the cooling water remains constant, the velocities increase, and heat transfer

12 230 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 6 Properties of GlidCop AL-25 as a function of temperature Tempera- Alpha Heat capacity Yield strength Ultimate Young s mod- Density Thermal conduc- Poisson s ture ( C) ( C 1 )( 10 6 ) (J kg 1 C 1 ) (MPa) strength (MPa) ulus (GPa) (mg m 3 ) tivity ratio (W m 1 C 1 ) Equation T T T T T T T T T T T T T T

13 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 5. Electrical resistivity of Glidcop AL-15 as a function of temperature. coefficients go higher, and the heat flux to the tube increases. The effect on the maximum temperature of the Glidcop is not very high. Table 8 gives the thermal hydraulic parameters after 3 FPY of operation. The average increase in the ohmic heating in the middle straight section is Fig. 6. Resistivity change due to nuclear transmutation in copper after 6 FPY. 19.5% and in the transition regions 9.25%. The upper and lower straight sections are unaffected. Water mass flow rate increases by 11.8% and the velocities likewise. The maximum Glidcop temperature goes from to C, an increase Table 7 Thermal hydraulics parameters of center-post prior to irradiation Upper straight Upper transition Middle straight Lower transition Lower straight Water mass flow (kg s 1 ) Water inlet temperature ( C) Water outlet temperature ( C) Water velocity (m s 1 ) Cu resistivity (ohm.cm 10 6 ) Ohmic power (MW) Heat flux to tubes (W cm 2 ) Heat trans. coefficient (W cm 2 K 1 ) Average Cu conductivity (W cm 1 K 1 ) Cu maximum temperature ( C) Water pressure drop (MPa)

14 232 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 8 Thermal hydraulics parameters of center-post after 3 FPY Upper straight Upper transition Middle straight Lower transition Lower straight Water mass flow (kg s 1 ) Water inlet temperature ( C) Water outlet temperature ( C) Water velocity (m s 1 ) Ohmic power (MW) Heat flux to tubes (W cm 2 ) Heat transfer coefficient (W cm 2 K 1 ) Average Cu conductivity (W cm 1 K 1 ) Cu maximum temperature ( C) Water pressure drop (MPa) of 5%. The major effects are on water pressure drop and pumping power which increase by 23 and 37%, respectively. Nevertheless, pumping power at 56.4 kw is still very reasonable. The sum of ohmic heating and nuclear heating in the center-post is 81.9 MW at the start of operation and MW after 3 FPY. After 6 FPY of operation a similar increase in resistivity will occur making the total ohmic plus nuclear heating equal to MW. The water mass flow rate will go up to 306 kg s 1, the total pressure drop to 0.25 MPa and the pumping power to 77.3 kw. At that point the maximum water velocity will be 9.2 m s 1 at the mid-plane. Table 9 summarizes the thermal hydraulic parameters of the center-post Structural consideration There are two major structural concerns on the center-post. They are the compressive forces due to the flowing current, and the tensile strength exerted on the center-post due to expansion forces from the TF outboard back-legs. Since Glidcop becomes brittle after a short exposure to radiation, it is prudent to alleviate any tensile loads if possible. Fortuitously, in the case of VNS it was possible to isolate the expansion force on the center-post by designing a sliding electrical connection between the center-post and the outboard TF. This will be discussed in the following sections Compression forces Any electrical conductor is subjected to compression forces due to the self-induced magnetic field produced by the current flowing through it. The magnetic pressure is given as: Pressure=B 2 / 0 where B is the peak magnetic field at the surface of the conductor and 0 is the permeability of air in vacuum. At a current of 7.16 MA and a radius of 16.7 cm, the peak field is 8.5 T. The pressure on the center-post is then equal to 28.9 MPa. Table 10 gives the properties of irradiated Glidcop alloys after 3 and 15 dpa. It can be seen that the yield strength at room temperature stays essentially constant at 340 MPa after irradiation, and from Table 6 it is seen that the yield strength at 200 C is 239 MPa. Thus, it is concluded that there is a margin of a factor of 8 on compression Expansion forces The total force acting on the top of the bell jar, between the center-post r c and the return conductor r r is: 2 r r dr F={I(A)} r 10 7 (Newtons) r c ={I(A)} 2 ln r r r c 10 7 =13.19 MN where I is the current in amps and r is in meters. The resulting pressure is MPa.

15 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 9 Summary of the thermal hydraulics parameters of the VNS center-post At start of operation After 3 FPY After 6 FPY Total power dissipated (MW) Water mass flow rate (kg s 1 ) Maximum Cu temperature ( C) Coolant pressure drop (MPa) Pumping power (kw) To estimate the deflection at each end of the center-post, the correlation for a built in disc has been used, free in the middle and with a uniform pressure of MPa as shown in Fig. 7. The upper plate is 0.3 m thick and its plate constant D is 326 MNm. The deflection at the center-post is 3.2 mm. Since the bottom end is fixed, the total deflection at the top will be 6.4 mm. The moment at the plate outer radius, which is built in, is 24.2 MNm m 1. To reduce the stress at the corner, the thickness will be increased to 1.0 m. The bending stress is then: s bend = 6M =145 MPa 2 t O erturning moments Overturning moments appear due to the J B Lorentz forces in the return legs of the TF coils coming from the interaction of the poloidal field with the electric current in the TF coils. During reactor operation when the plasma current is in place, the poloidal fields are low. However, when a disruption occurs and there is no plasma current to reduce the poloidal field, these J B forces are the highest. Their manifestation is a torque on the upper and lower lid of the bell jar which is ultimately applied to the center-post, unless the sliding electrical connection can accommodate a rotational displacement without imparting the load to the center-post. By calculating the magnetic field contributions by the individual poloidal field coils at the TF coil return legs (i.e. the upper and lower lids of the bell jar), summing them, crossing the resulting field with the TF coil current and integrating the resulting moments the total torque on the lids can be obtained, which is equal to 87.2 MNm. If this torque is applied to the cylindrical portion of the bell jar using only the material between the test modules access ports, the total twist is equal to radians, or degrees. Assuming rotation can only take place at the top, the total twist will be twice as high, or degrees. At the omega bellows this is a rotation of 0.14 mm which can be easily accommodated. The conclusion is that a set of omega bellows enveloping a quantity of mercury can possibly be designed to sustain a deflection of 6.4 mm and a twist of 0.14 mm. The maximum shear stress on the cylindrical portion of the bell jar is 10 MPa Sliding electrical connection As mentioned earlier, in order to relieve the center-post from tensile forces, a sliding electrical connection will be provided at the upper interface between the center-post and the outboard TF legs. In this case, using the bell jar concept, the connection is made between a thick washer and the center-post. The sliding connection is provided by means of a volume of mercury trapped between two (omega) bellows. Fig. 8 is a side view of the connection scheme. The single turn omega bellows are welded to flanges which are sealed with metallic O rings to mating flanges on the center-post on one end and to the connecting washer on the other. Any mercury leaks from this system will be contained by the vacuum seal below the omega bellows assembly. Electrical current flows from the washer to the center-post through a 0.5 cm layer of mercury. Under load, the washer moves up 6.4 mm (see previous section) and the omega bellows will permit such a deflection without imparting a tensile load on the center-post.

16 234 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 10 Tensile properties of irradiated GlidCop alloys Material Condition Tensile stress Yield stress Elongation (%) Red. in area (%) MPa KSI MPa KSI GlidCop C15720 Control dpa dpa C15760 Control dpa dpa The resistivity of mercury is ohm cm, the gap is 0.5 cm, the contact area is 0.88 m 2 and the current is MA. The ohmic heating is 0.3 MW and can be dissipated by conduction to the cooled surfaces of the center-post and the washer. Some question arises as to the compatibility of Hg and Cu and the wetting characteristics. It might be foreseen that plating may be needed to prevent corrosion and promote wetting. This analysis is reserved for future work First wall for center-post To eliminate surface heating on the centerpost, it will be equipped with a first wall (FW). The material for the FW will most likely be 316 LN or 304 stainless steel. Fig. 9 shows a possible design for the FW. It consists of rectangular channels sandwiched between two flat sheets. The same number of channels runs the full height of the FW, but since the circumference of the center-post changes, the aspect ratio of the channel changes. Section A A shows the geometry at the center-post extremities, while Section B B is the geometry at the mid-plane. For the present analysis, it is assumed that 50% of the power leaving the plasma in the form of radiant energy and lost particles will end up on the chamber wall. This power, which consists of 20% of the fusion power (39 MW) and 100% of the neutral beam power (21 MW), is estimated to be 28.8 MW. Of which 14.4 MW will be incident on the FW. The center-post FW area is 9.35 m 2 and the outboard (OB) FW is 46.5 m 2. The energy incident on the center-post FW is 2.41 MW, giving an average heat flux of 0.26 MW m 2. Using the nuclear peaking factor of 1.49, one gets the peak heat flux of 0.38 MW m 2. Assuming a2mmthick FW consisting of a 1 mm thick front sheet and a1mmthick coolant channel thickness, 38.4 W cm 2 surface heating and 10 W cm 3 of nuclear heating, a thermal stress thfw =106.4 Mpa would be obtained. The circumference of the FW at mid-plane is cm and if assuming channels of 0.5 cm wide and 1.5 cm deep, there will be 163 channels if the channel wall thickness is 1 mm. At the extremities where the circumference is 201 cm, the channels will be 1.03 cm wide and 1.5 cm deep. For a total heating of 3.11 MW of which 0.7 MW is nuclear heating, and using a T of 30 C, a mass flow rate of kg s 1 and a velocity of 2.5 m s 1 at the mid-plane will be obtained. The maximum temperature of the FW surface is only 152 C Center-post FW fabrication Fig. 9 shows that the FW is made from two halves and is assembled around the center-post clam-shell-wise. There are manifolds at each end of each half, which means there will be two supply and two return headers for the FW. Coolant comes in at the top and exists at the bottom.

17 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 7. Model used for determining deflection and stresses in upper lid of the bell jar. The two halves of the FW will be identical in all respects. Each of the 163 tubes will also be identical. The tubes can be explosively shaped inside a form giving them the varying rectangular crosssections as described in Fig. 9. They are then bent to conform to the longitudinal shape of the center-post. The inner and outer sheets between which the channels are sandwiched are formed on a mandrel. The channels are then assembled onto the outer sheet, and continuously spot welded along their whole length. After all the channels are assembled, the inner sheet is added and continuously spot welded as in the case of the outer sheet. The two halves of the FW are then assembled clam-shell-wise and welded together as shown in Fig. 9. Fig. 8. Design of a sliding electrical connector using mercury. Fig. 9. First wall design for the center-post.

18 236 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Table 11 Summary of divertor thermo-mechanical analysis Total power to divertors (MW) 15 Inlet coolant temperature ( C) 40 Outlet coolant temperature ( C) 120 Water mass flow rate (both sides) (kg s 1 ) 92.6 Average heat flux (MW m 2 ) 1.3 Peak heat flux (MW m 2 ) 5.2 Maximum coolant velocity (m s 1 ) 5.0 Maximum Cu temperature ( C) 325 Maximum Be temperature ( C) 416 Peak Von-Misses stress in Cu (MPa) 143 Peak Von-Misses stress in Be (MPa) 312 Fig. 10. Design of the divertor plate The di ertor design The VNS divertor consists of two assemblies, one on each end of the center-post as shown in Fig. 1. The divertors are shown as cones sloping away from the center-post at 45 angles to the horizontal. The exact heat flux to the divertor has yet to be determined, but for the present, it is assumed that 50% of all the charged particle and NBI power will be incident on the divertors (see Section 2), which is estimated to be 14.4 MW. The total area of the two divertor plates is 11.6 m 2 Fig. 11. Cut view of a divertor coolant channel. which gives an average heat flux of 1.24 MW m 2. When a conservative peaking factor of 4 is added, the maximum heat flux is 5.0 MW m 2. The divertor plates are designed similar to the ITER limiters [37]. Coolant tubes imbedded in Cu blocks radiate outward from the center-post. The Cu blocks are attached to a cooled SS plate. The surfaces of the Cu blocks facing the plasma are covered with Be tiles 2 3 mm thick. Fig. 10 shows several views of the divertor plate. For the Cu blocks to cover the cone, they have to increase in width as they go to a larger radius. The coolant channel also gets larger in area. Sections A A and B B in Fig. 10 show how this can be accomplished. A 3D thermal hydraulics model has been constructed using a Be thickness of 0.2 cm, Cu thickness of 1.9 cm and a SS plate of 2 cm. To be consistent with the center-post cooling water, the temperature deferral is the same, water entering at 40 C and exiting at 120 C. The mass flow rate for each divertor is 46.3 kg s 1 and the maximum velocity at the peak of the heat flux is 5ms 1. The model, as shown in Fig. 11, represents one half of a coolant channel cut at a boundary of symmetry Y Z at X=0. The X Y surface where the Cu is attached to the SS is a zero stress condition, assuming that its temperature will be the same as that at the end of the annealing cycle. The X Z plane at Y=0 is assumed fixed through the Cu only, while the other end (the surface parallel to the X Z plane) is free to expand. Because the Cu blocks are discrete, the surface parallel to the Y Z plane at X=1.0 cm is also

19 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) free to expand. The Be surface is castellated with the maximum castellations not exceeding 1 cm 1 cm. The surface heat flux is 520 W cm 2. Temperature dependent material properties are used for the Be, Cu and 316 SS and the finite element thermal hydraulic and stress analysis code AN- SYS has been used. The results are summarized in Table 11. The maximum copper temperature is 325 C and the beryllium, 416 C. These values are well within the acceptable limits. Further, the Von- Misses maximum stress in the copper is 143 MPa. In taking 3/4 of yield strength at 325 C (see Table 6) one gets 132 MPa. However, the 3D analysis does not allow for yielding and this stress of 143 MPa in the copper is due to the singularity at the interface between the copper and beryllium. In the actual case, there will be a layer at the interface where there will be diffusion of the beryllium into the copper with somewhat intermediate coefficients of expansion. The same is true of the maximum Von-Misses stress in the beryllium of 312 MPa. The next phase of the study would be to do the 3D analysis using plastic deformation which mitigate the singularity at the interface. At that time it may be necessary to include an intermediate layer of material between the beryllium and copper to bring down the stress even further. The other question that needs clarification is how long can 2 mm of beryllium last before it becomes eroded. Present experiments at Los Alamos National Laboratory show that beryllium can be plasma sprayed achieving a very high density and properties close to 100% dense beryllium. This may open the possibility of in-situ rebuilding the beryllium coating on the divertors. For performing in-situ rebuilding of the beryllium surface, the plasma re-coating equipment must be placed inside the chamber. This can be done through one of the test module ports which will be specially equipped for the purpose. Rails can be set up within the chamber which will allow a plasma re-coating head to circumvent the centerpost while replacing the beryllium layer on the divertor plates. Should that not be possible, then a design requirement will be to conceive a system whereby the divertor plates can be replaced in a Fig. 12. Figure supplied by SCM Corporation for manufacturing the center-post. short enough time as not to appreciably impact the availability of the device Center-post fabrication The primary option for the center-post is to make it from Glidcop C15715 or C Several meetings have taken place with representatives of SCM Metal Products, Incorporated, to discuss the fabrication process. Three possible methods have been proposed and each will be discussed separately. Glidcop is made with a patented powder metallurgy process by SCM Metal Products, Incorporated, using the internal oxidation technique. The process begins with melting a dilute solid solution of Cu and Al. It is then atomized into powder and placed in forms made from Cu according to the required shapes and the form sealed. During heating the Al in the alloy is selectively oxidized Fig. 13. Another SCM Corporation figure for a second approach.

20 238 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 14. Another SCM Corporation figure for a third approach. In Section 3.2, the progressive increase in the Glidcop resistivity due to radiation transmutation was discussed. It was shown that the overall cenin-situ. The powder is then consolidated into fully dense shapes by one of several conventional metal working processes. The forms can have pre-positioned tubes made of either SS or Cu, which after consolidation become an integral part of the assembly. The present oven facilities are limited in the sizes of parts they can produce. To fabricate the whole center-post as a complete unit, the oven has to be much larger than what presently is available. However, it has been indicated that there is no fundamental reason why the centerpost cannot be fabricated as a complete unit if an oven large enough to accommodate it is available. Fig. 12 is a crude picture from SCM of the complete center-post. In this proposal, the tubes are SS, but they also could be made of normal Cu or Glidcop. The tubes are located in place by means of Cu positioning fixtures and then a Cu form is placed to surround the center-post. The form is then filled with the DS Cu Glidcop alloy, and the form is sealed all around, including the tubes. The whole form is then subjected to heat and high pressure and is consolidated into a homogeneous unit. Machining off the form is the final step in the process, however, in the case of the center-post, this may not be necessary. In the second method shown in Fig. 13, a central core is extruded and the coolant channels are gun drilled. The upper and lower transition region are produced separately and jointed to the core with a hot-isostatic-pressure (HIP) technique. It was decided that connecting channels from the transition regions to the core made this method impractical. The third method shown in Fig. 14 shows the center-post made from three sections which are diffusion bonded together. This method has the advantage of smaller production pieces. However, the diffusion bonding will most likely require an oven large enough to contain the center-post anyway. Some discussion centered on the oven and how much it would cost to produce. Because the shape of the center-post is long and thin, the oven can be made from seamless thick wall pipe capable of sustaining the high pressure (approximately 10 MPa) needed for consolidation. Such an oven would cost on the order of 5 10 M$. Another method for producing the center-post from conventional wrought Cu or Glidcop is shown in Fig. 15. In this method, the center-post is divided into 8 or 12 wedges (there are 12 in the figure). If the wedges are made from wrought Cu such as Cu Ni Be, the wedges are machined with the grooves as shown in the figure. The wedges are then assembled and either diffusion bonded or brazed together to form the complete center-post. The grooves in the wedges are optimized for that particular section of center-post, and they propagate the whole length, varying in width to control the coolant fraction Center-post replacement Fig. 15. Alternate scheme for manufacturing center-post using wrought copper.

21 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 16. Center-post replacement step No. 1. ter-post resistivity increased by 25% after 6 FPY of operation, and the coolant pumping power increased by almost 90%. Whether due to thermal hydraulic, activation or structural reasons, there will come a time when the center-post will have to be replaced. This section describes the steps needed to replace the center-post and estimates the required time. A significant aspect of the center-post replacement is the risk of exposing surrounding systems to activated components. In this design, avoidance of such exposure is insured by removing components from the bottom of the device and placing them directly into shielded casks. Thus, after the peripheral components are disconnected and removed, the activated center-post is lowered into a cask which can be equipped with a convective gas circulation system. Ultimately, the used center-post can be stored under water in a pool storage area. Figs give a progression of steps needed to replace the center-post and Table 12 gives the estimated times required. These times have been arrived at by comparing similar tasks proposed for the maintenance of ITER. Where no such similar tasks were available, an estimate was made of how long it would take to perform hands on and a factor of 8 10 was applied to reflect a remote operation. The approximate time for removing the centerpost from the reactor, placing it into a cask and removing the cask to a storage area is 98 h. Preparing the chamber for a new center-post is estimated at 40 h. This step primarily entails reconditioning surfaces that were disconnected and cleaning up and preparing new weld interfaces. Inserting and securing a new center-post, and sealing and testing the assembly is assumed to take three times as long as removing the centerpost, or 300 h. The total time needed is 438 h exclusive of the cool-down period which is estimated to be about 24 h. Assuming two shifts of 40 h per week, the time needed is 6 weeks. If replacement takes place every 6 FPY, the impact on availability will be minimal, only 1 week per FPY. If the center-post is replaced every 3 FPY, the impact becomes 2 weeks per FPY Other maintenance requirements The lifetime of the divertor plates is difficult to estimate at this time, especially in view of the fact that in-situ re-coating of the Be may be possible.

22 240 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 17. Center-post replacement step No. 2. Replacement of the divertor plates can be made at the same time as the center-post. The additional steps needed will add 3 4 more weeks to the maintenance downtime. Removing the conical divertor shield plugs can also be done through the bottom of the reactor. The divertor plates can then be replaced in a hot cell and new plates mounted onto the shield plugs which should be able to be reused. The outboard FW components may also have to be replaced. After 6 FPY they will have accumulated on average 6.18 MW m 2 and at the peak, 6.96 MW m 2. This is equivalent to 69 dpa. Replacing the outboard FW components will add 8 weeks to the maintenance downtime. The cumulative time needed to replace the center-post, divertor plates and the FW will be 18 weeks. Even at this downtime, the impact on availability will be only 3 weeks per FPY, or 6% Power supplies The total power requirement of VNS will be 200 MW but roughly half of that is needed for the TF coils. The main requirement of the TF power is the very high current of 7.2 MA. In the present design, the power is taken through a single step-down 400 kv/22 kv, 240 MVA transformer. At 22 kv, the power is distributed through six radial feeder breakers to 42 MVA step-down regulator transformers, with on-load tap changers providing a 40% range of output voltage variation. Each regulating transformer supplies two thyristor regulators, which in turn, each control a pair of converter modules. The outputs from the converters are connected to the TF coils by a pair of 300 ka conductors for each of the 24 busses. The output current from each pair of 300 ka modules is monitored with a DC current transformer whose output is summed with that from the other 23 to give a total current signal. The signal is used to control the outer loop servos, which in turn sets the reference for all the individual inner loop servos associated with each thyristor controller. Controller out-of-range and low power factor signals are derived to initiate appropriate tap changes simultaneously on all connected regulators. In this way, the total DC load is maintained in balance on all connected converters, and phase delay angle excursions are limited to within that necessary to bridge a tap voltage of the regulating transformers.

23 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 18. Center-post replacement step No Nuclear performance This section summarizes the nuclear analysis investigation for the reference (39 MW) ST VNS facility. A two-dimensional neutronics model, which simulates the ST VNS configuration as depicted in Fig. 1, was employed for the analysis. Neutronics calculations were performed with the Los Alamos developed 2D discrete ordinates neutron and photon transport code, TWODANT [38]. The nuclear data library employed is the ENDF/B-V based, multigroup neutron (30 groups) and -ray (12 groups) coupled transport library, MATXS5, processed by MacFarlane of Los Alamos National Laboratory [39]. Neutron activation calculations were performed using the activation calculation code, REAC3 [40]. The activation cross section and decay data libraries are those derived from USACT93 and maintained by Mann of Westinghouse Hanford Company [41] Neutronics model The principal parameters of the ST based reference VNS device as they affect the neutronics are summarized in Table 13. The schematic of the two dimensional neutronics model is shown in Fig. 22. In order to provide a geometry to model the source profile as provided and to model the divertor as seen in Fig. 22, a fine (2.5 cm 2.5 cm) mesh was used. The region-wise geometry and material specifications are given in Table Neutron source The 14 MeV neutron source spatial distribution in the plasma zone is shown in Fig. 23. The intensity factors used in the TWODANT calculations are given in Table 15. The source normalization used in the calculation was ns 1 in the entire plasma zone accounting for a 39 MW fusion power (or 31.2 MW neutron power) Neutron fluxes Fig. 24 shows a contour plot of the 14 MeV neutron fluxes (in log 10 scale) in the reference ST VNS device. As seen in this figure, the neutron fluxes in the various source regions within the plasma zone are well represented. The attenuation of neutron fluxes in the test section is also shown

24 242 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 19. Center-post replacement step No. 4. to be not as effective as in the other SS316 and water, or copper and water zones. Fig. 25 shows the mid-plane and the device top (3.85 m from the mid-plane) radial neutron flux profiles. Note that as seen in Fig. 25, the neutron flux may drop by more than ten orders of magnitude vertically along the center-post. However, at the divertor location, the reduction is only about eight orders of magnitude because of less materials shown in that region Neutron wall loading Table 16 summarizes the results of the neutron wall loading calculations. As can be seen from Table 16, the average neutron wall loading in the test section meets the design criteria of 1 MW m 2 [1]. The axial neutron wall loading profile extending to the bottom of the divertor is shown in Fig. 26. The neutron wall loading along the diagonal of the wedge shape divertor is shown in Fig. 27. The maximum neutron wall loading at the bottom of the divertor (20 cm flat section) is 0.50 MW m 2, while the neutron wall loading along the diagonal ranges from 0.4 to 0.2 MW m Nuclear heating Fig. 28 shows the mid-plane radial heating profile. The maximum heating rate occurs at the inboard coil is 10 W cm 3, and it occurs at the coil-inboard wall interface. Table 17 gives the nuclear heating rates for the major components Atomic displacement and gas production rate The axial atomic displacement rate profiles at the inboard wall and center-post are shown in Fig. 29. The maximum atomic displacement rate for the stainless steel occurs at the inboard wall and is 8.8 dpa per year. The maximum atomic displacement rate for copper in the center-post is 8.2 dpa per year, and occurs at the coil-wall interface. The maximum atomic displacement rate at the center of the center-post is 3.6 dpa per year. The axial atomic displacement rate profile for the outboard wall is shown in Fig. 30. The maximum atomic displacement rate for the vanadium alloy first wall at the test section is approximately 11 dpa per year. The maximum atomic displace-

25 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 20. Center-post replacement step No. 5. ment rate for the outboard shield stainless steel wall is approximately 6 dpa per year. A detailed contour plot of the atomic displacement rates in the center-post is shown in Fig. 31. As shown in this figure, the atomic displacement rate in copper ranges from 3 to 9 dpa per year at the mid-plane due to the flux attenuation. It also varies significantly on the vertical direction due to the wall loading distribution. Note that it is about 2 m in height for a factor of 10 reduction in the atomic displacement. The helium to dpa ratio for SS316 is 12 at the mid-plane first wall. The hydrogen to dpa ratio is about 40. The helium to dpa ratio for copper at the center-post and first wall interface is 26 and increases to 48 at the center of the center-post where the atomic displacement is lower. The hydrogen to dpa ratio is 14 at the interface and maintains about the same at the center Copper transmutation in center-post Transmutation of copper due to nuclear interactions with neutrons has been identified as an important design issue for the normal conducting TF coil in a neutron producing experimental device. The major impact is the increase of resistivity in copper because of the attending transmutation products. As a consequence, the ohmic heating power will increase and the operating thermal-hydraulic and temperature characteristics in the copper coil will change. Activation analysis of the center-post copper reveals that the dominant transmutation products are Ni, Zn, Co, and Li. Nickel isotopes are produced due to the following nuclear reactions: 63 Cu(n,p) 63 Ni, 63 Cu(n, ) 64 Cu( +) 64 Ni, 63 Cu(n, 2n) 62 Cu( +) 62 Ni, and 65 Cu(n,2n) 64 Cu( +) 64 Ni. Zinc isotopes are generated via: 63 Cu(n, ) 64 Cu( ) 64 Zn, 65 Cu(n,2n) 64 Cu( ) 64 Zn, and 65 Cu(n, ) 66 Cu( ) 66 Zu. Cobalt isotopes are produced by 63 Cu(n, ) 60 Co and 63 Cu(n,n ) 59 Co reactions. Lithium isotope is generated in Glidcop because of the 10 B(n, ) 7 Li in the impurity element boron found in that material. Of these transmutation products, Ni and Zn were found to be responsible for the increase of electrical resistivity mainly because of their high specific resistivity coefficients, and ohm.m per

26 244 E.T. Cheng et al. / Fusion Engineering and Design 38 (1998) Fig. 21. Center-post replacement step No. 6. atom percent, respectively. The highest transmutation rates of copper into nickel and zinc occur at the interface between the center-post and the inboard first wall at the mid-plane. The are 0.17 and 0.08 atom percent per year, respectively, for nickel and zinc. The corresponding resistivity increase rate is estimated to be 8.3% per year. Using the calculated transmutation rates and the specific resistivity increase coefficients, the spatial resistivity increase rates in copper can be estimated. Fig. 6 displays the contour of the resistivity change in the center-post after 6 FPY (see Section 3.2) Decay heat Fig. 32 displays the decay power density for SS316 and Cu at various center-post locations, as a function of time after 3 FPY exposure. The decay power density at shutdown is about 2 3% for Cu and SS316, as shown in Fig. 32, and then drops to about 0.6 and 0.2%, respectively, in a few days. At the top of the center-post, due to the neutron attenuation, the decay heat is more than ten orders of magnitude lower than that at the mid-plane, as also shown in Fig. 32. Fig. 33 shows the total decay power in the center-post component. Note that the total decay power at shutdown is 25 kw, as shown in Fig. 33, which is about 1.6% of the total nuclear heating in the center-post. Compared with the ohmic heating in the center-post, which ranges from 82 MW at the beginning of lifetime to 92 MW after 3 FPY exposure, the decay heat is negligible. One of the important factors affecting the availability of the VNS device is the cooling time required before the cooling pipe can be disconnected for the replacement of the center-post to occur. This is because the temperature rise due to the decay heat in the few days (which is about 4 days as discussed in Section 3.8) when the centerpost component is without the coolant can not be too excessive. To obtain a preliminary estimate, calculations were performed assuming conservatively that the temperature rise in the center-post is due to adiabatic heating by the decay heat. The results are displayed in Fig. 34 and Fig. 35 for 3 FPY and 6 FPY exposures, respectively. As shown in Fig. 34, the 4-day adiabatic temperature rise without the coolant is about 850 C when no cooling time is allowed. However, this